The present invention relates generally to improving the operating characteristics of on-line nuclear reactors. More specifically, it relates to improving the useful life of a reactor which has been in operation by limiting or reducing further stress corrosion cracking of components of a boiling water reactor or other reactor components exposed to high-temperature water.
Boiling water reactor components are known to undergo stress corrosion cracking. Stress corrosion cracking is a phenomenon which occurs in apparatus exposed to high temperature and, accordingly, high-pressure water as well as at lower temperatures. The stress arises from differences in thermal expansion, the high pressure needed for the containment, and other sources including residual stress from welding, cold work and other asymmetric treatments. In addition to the stress, other conditions including sensitization of the metal and water chemistry influence the sensitivity to stress corrosion cracking (SCC). This type of corrosion has been widely studied and a number of papers have been written concerning it.
Among them are:
1) F. P. Ford, "Stress Corrosion Cracking", in Corrosion Processes, edited by R. N. Parkins, Applied Science Publishers, New York, 1982, p. 271. PA1 2) J. N. Kass and R. L. Cowan, "Hydrogen Water Chemistry Technology for BWRs", in Proc. 2nd Int. Conf on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Monterey, Calif., 1985, p. 211. PA1 3) M. E. Indig, B. M. Gordon, R. B. Davis and J. E. Weber, "Evaluation of In-Reactor Intergranular Stress" in Proc. 2nd Int. Conf on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Monterey, Calif., 1985, p. 411. PA1 4) L. G. Ljungberg, D. Cubicciotti and M. Trolle, "Materials Behavior in Alternate (Hydrogen) Water Chemistry in the Ringhals-1 Boiling Water Reactor", Corrosion, Vol.42, (1986) p. 263. PA1 5) L. W. Niedrach and W. H. Stoddard, "Corrosion Potentials and Corrosion Behavior of AISI304 Stainless Steel In High Temperature Water Containing Both Dissolved Hydrogen and Oxygen", Corrosion, Vol. 42, No. 12 (1986) page 696. PA1 (6) H. Ocken, C. C. Lin, and D. H. Lister, "Thin Films to Impede the Incorporation of Radio Nucleides in Austenetic Stainless Steels", Thin Solid Films, Vol. 171 (1989) pages 323-334. PA1 (7) G. P. Chernova, T. A. Fedosceva, L. P. Kornienko, and N. D. Tomashov, "Increasing Passivation Ability and Corrosion Resistance of Stainless Steel by Surface Alloying with Palladium", Prot. Met. (Eng. Transl.) 17 (1981) page 406.
It is well documented that stress corrosion cracking occurs at higher rates when oxygen is present in the reactor water in higher concentrations.
As explained in these and other articles, efforts have been made to lower the stress corrosion cracking in boiling water reactor piping by lowering oxygen levels in the cooling water through hydrogen injection to achieve higher concentrations than normally present in the water as a result of radiological decomposition. It has been found that varying amounts of hydrogen have been required to reduce oxygen levels sufficiently to achieve and reliably maintain the critical potential required for protection from the SCC in the high temperature, high pressure water. Accordingly, the problem of stress corrosion cracking of stainless steel components, including piping of boiling water reactors, has remained a significant problem. The present invention is aimed at reducing the amount of H.sub.2 required as well as at facilitating reliable maintenance of the corrosion potential below a critical value of -230 to -300 mV vs. the standard hydrogen electrode (SHE) at which SCC is markedly reduced or even eliminated as indicated in references 3 and 4.
Two additional papers deal with the formation of noble metal deposits on reactor piping and other containment structures and they are as follows:
The first of these articles deals with the use of metal deposits and other treatments and deposits to reduce the build-up of radioactivity in components of the circulatory system of a nuclear reactor that are in contact with the coolant.
The second of these articles deals with the electrochemical behavior and general corrosion resistance of stainless steel as distinct from stress corrosion cracking.
One of the by-products of the presence of oxygen in the hot water in contact with the interior of reactor piping and other reactor components is the formation of a surface coating of mixed oxides. The surface coating forms directly on the surface of stainless steel and is a mixture of oxides of the metals of the stainless steel including chromium, iron, and nickel. The surface coating increases in thickness from the time the stainless steel surface is first exposed to the oxygen containing high pressure, high temperature water. The present invention is directed particularly toward reducing the stress corrosion cracking in reactor piping and in other reactor components for reactors which have been on line and have operated for a period of time and which consequently have an interior surface coating of mixed oxides.